IR 05000293/1999002
| ML20206U757 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 05/17/1999 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20206U754 | List: |
| References | |
| 50-293-99-02, 50-293-99-2, NUDOCS 9905260004 | |
| Download: ML20206U757 (18) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
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License No.:
DPR-35 i
Report No.:
99-02 l
Docket No.:
50-293 Licensee:
BEC Energy 800 Boylston Street Boston, Massachusetts 02199 Facility:
Pilgrim Nuclear Power Station inspection Period:
March 8,1999, through April 18,1999 l
Inspectors:
R. Laura, Senior Resident inspector R. Arrighi, Resident inspector Approved by:
C. Anderson, Chief Projects Branch 5
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Division of Reactor Projects s
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9905260004 990517 P
PDR ADOCK 05000293
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EXECUTIVE SUMMARY
Pilgrim Nuclear Power Station NRC Inspection Report 50-293/99-02
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This routine inspection included aspects of licensee operations, engineering, maintenance, and plant support. The report covers resident inspection for the period of March 8,1999, through April 18,1999.
Operations The nuclear plant operator identified several equipment problems during a tour of the
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reactor building; and appeared to be very knowledgeable and experienced. (Section 01.1)
Operators incorrectly moved several control rods during a planned down power due to
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inattention-to-detail and poor communication. The licensee's immediate corrective actions to address this issue were determined to be good. This Severity Level IV violation is being treated as a Non-Cited Violation, consistent with Appendix C of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as PR 99.9108 (NCV 50-293/99-02-01). (Section O1.2)
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Operators promptly declared main steam isolation valve AO-220-2C inoperable due to a
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slow close test failure and followed technical specifications and other administrative requirements. Subsequently, operators responded well to a small power excursion when operating at 85% reactor power with one steam lino isolated. (Section O6.1)
Maintenang Overall good pre-job briefs and procedure adherence was displayed during routine
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maintenance and surveillance activities. (Section M1.1)
Planned corrective maintenance to replace a fuel injector and high pressure tube on the
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station blackout (SBO) emergency diesel generator (EDG) was successfully completed by competent mechanics. Good attention te detail was evident by thorough cleaning of carbon residue where the injeclor seals insert inside the cylinder head. (Section M1.1)
Two different maintenance work package quality issues resulted in wrong parts being
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sent to the job site for installation in safety related equipment. This was identified by maintenance field workers but reflected errors by planning personnel. In the first example, a work planner error resulted in sending a DC electrical relay coil instead of an AC relay. In the second example, a planner twice ordered the wrong top works for a hand operated valve in the EDG fuel oil transfer system. This Severity Level IV violation is being treated as a Non-Cited Violation, consistent with Appendix C of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as PR 99.0620 and PR 99.0626 (NCV 50-293/99-02-02). (Section M1.1)
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i Executive Summary (cont'd)
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The apparent cause evaluation for the wrong valve parts did not fully evaluate all aspects
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of how the work planner ordered the wrong parts twice. (Section M1.1)
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Extensive planning was evident for RFO12 with an emphasis on minimizing shutdown risk. Adequate administrative controls were in place to control in vessel activities such as repairing a stuck control rod drive and maintaining the reactor fuel offload rate within the capabilities of the decay heat removal systems. (Section M2.1)
Quality assurance prepared an RFO12 inspection plan partly based on risk significance
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which included some nonsafety related systems. (Section M2.1)
Enaineerina The plant design changes (PDC) reviewed clearly identified those controlled documents
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and training activities that were required to be updated prior to turnover of the component / system to operation. The PDCs were properly implemented in accordance with NRC and licensee requirements. (E2.1)
Plant Sucoort Through observation of ongoing activities, the inspector concluded that the licensee's
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radiological controls for routine activities were properly being implemented. (R1)
Security measures for personnel access and security measures for temporary structures
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were properly being implemented. (S1)
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TABLE OF CONTENTS
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EXECUTIVE SUMMARY....................................................... ii Sum mary of Plant Status...................................................... 1
Conduct of Operations............................................ 1 01.1 ' General Comments........................................ 1 01.2 Control Rod Pull Sheet.................................... 2
Operator Knowledge and Performance............................... 3 06.1 Main Steam isolation Valve (MSIV) Operability issue..............
II. MAI NTE NANCE........................................................... 4 M1 Conduct of Maintenance.......................................... 4 M1.1 General Maintenance....................................... 4 M2 Maintenance and Material Condition of Facilities and Equipment..........
M2.1 Refueling Outage No.12 (RFO12) Preparations................ 7 M8 Miscellaneous Maintenance issues.................................. 8 M8.1 (Closed) LER 50-293/98-25: Setpoint of Target Rock Relief Valve Found Out of Tolerance During Testing........................ 8 M8.2 (Closed) LER 50-293/98-26: High Pressure Coolant injection (HPCl)
System Declared inoperable Due to Failed Power Inverter.........
M8.3 (Closed) LER 50-293/98-27: Switchgear Room Door inoperable for Tomado Analysis.......................................... 9 Ill. ENGI N EERI NG........................................................ 9 E2 Engineerin0 Support of Facilities and Equipment....................... 9 E2.1 - Plant Design Change Review................................ 9 IV. PLANT S U PPORT....................................................... 10 R1 Radiological Protection and Chemistry (RP&C) Controls................ 10 S1 Conduct of Security and Safeguards Activities....................... 10 V. MANAGEM ENT M EETINGS....................... '......................... 11 X1 Exit Meeting Summary........................................... 11 X3 Management Meeting Summary................................... 11 ATTACHMENTS
- Attachment 1 - Inspection Procedures Used
- Items Opened, Closed, and Updated
- Ust of Acronyms Used
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REPORT DETAILS
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Summarv of Plant Status Pilgrim Nuclear Power Station (PNPS) began the period at approximately 100 percent power.
On March 25,1999, power was reduced to 50 percent to perform a main condenser backwash, control rod pattem adjustment, and valve testing. During main steam isolation valve (MSIV)
testing, the *C" MSIV did not fully close during slow closure testing. Operators declared the valve inoperable and restored reactor power to 80 percent on March 29. Power was raised to 85 percent on April 5 then lowered back to 80 percent due to steam pressure oscillations.
Conduct of Operations 01.1 General Comments (71707)
Using Inspection Procedure 71707, the inspector conducted frequent reviews of ongoing plant operations, including observations of operator evaluations in the control room; walk-downs of the main control boards; tours of radiological controlled areas; and observations of several management planning meetings. The inspector observed that proper control room staffing was maintained and observed shift briefings and turnovers were well conducted with good discussion on compensatory measures and degraded equipment.
During deep back shift inspection, the inspector accompanied the reactor building tour operator during his rounds. The nuclear plant operator (NPO) was very knowledgeable and experienced. The NPO identified problems during the tour including a failed emergency light for the standby liquid control (SBLC) area. Corrective action was initiated by the issuance of a problem report. The inspector identified no problems and determined that the NPO completed the tour duties in a professional and diligent i
manner.
j During routine tours of the plant areas and control room, the inspector identified a few deficiencies which were reported to control room operators for resolution. A loose hanger was observed on a SBLC test line due to loose fasteners. Operators initiated a work request tag for corrective action. Also, a refuel floor radiation monitor recorder (i.e.,
RR-1705-21) on a back control room panel was not recording data. The inspector informed the reactor operator who replaced the red pen which had run out of ink. Lastly, the inspector observed two small leaks in the roof of the reactor building that was leaking
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rain water onto the refueling floor, in response, operators initiated a problem report to
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evaluate and correct the condition. Operators determined that the leaks were i
sufficiently small thus the operability of secondary containment was not in question.
Corrective actions were planned to repair the roof leaks. Lastly, the inspector noted that a section of the control room high efficience air filtration (CRHEAF) system flexible ducting was damaged. A small hole was identified in the discharge duct work of the "A" CRHEAF system. This discrepance was discussed with the nuclear watch engineer and problem report PR99.9078 generated. The licensee performed the quarterly CRHEAF surveillance test to verify system operability. Discussions with the licensee revealed that the damaged ducting was age related. These issues were not identified by the plant staff
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but were considered overall exceptions to the large number of problem reports initiated by the plant stsff.
01.2 Control Rod Pull Sheet a.
Inspection Scooe (71707)
A review was performed of the circumstances surrounding the inadvertent movement of several control rods during a planned reactor down power. The inspector also assessed l
the adequacy of the licensee's corrective actions that addressed this issue.
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Observations and Findinas On March 25 during the plant down power, operators incorrectly inserted six of the eight Group 16 control rods. Reactor operators were to insert control rods in reverse order of the pull sheets in accordance with the reactivity maneuver plan to decrease power to at or near the 67% flow control line. Control rod groups 16,11 and 10 were to remain at position 48 (skipped when inserting rods) to facilitate an efficient retum to 100 percent power to adhere to fuel warrantee limits.
During reactivity changes, operators follow a power maneuver plan that is developed by reactor engineering and provided to the Operations Manager for review. The evolution is performed by two licensed reactor operators with oversight from a dedicated on-shift reactivity manager. After inserting the group 17 control rods, reactor operators proceeded to insert group 16 rods. Six of the eight group 16 control rods were inserted from position 48 to position 26 before the error was identified by the on-shift reactivity manager. Control rod manipulations were stopped and problem report PR99.9108 generated. The licensee verified that acceptable thermal limits were maintained and repositioned the group 16 rods back to the desired rod pattem and continued with the planned down power.
The inspector attended the critique of the event, reviewed the reactivity maneuver plan, the rod withdrawal sheets and the licensee's corrective actions. The review of the rod withdrawal sheets revealed that the first rod in Group 16 was marked as skip with an arrow running down the side of the page for the remaining rods. The reactivity maneuver plan clearly called out that control rod group 16 was to remain at position 48. The licensee determined the cause of the event to be attributed to poor communications and inadequate attention to detail between reactor engineering and licensed operators. The licensee's immediate corrective actions included briefing all operators on the event and
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issuing standing order 99-01 requiring additional controls during rod maneuvers.
Although the control rods were not pulled in the desired order, this event posed no safety risk in that the rod withdrawal sheets were followed in the reverse order and therefore no thermal limits were exceeded. The failure to insert control rods in accordance with the.
rod withdrawal sheets is a violation of procedure 2.1.14, " Station Power Changes." This Severity Level IV violation is being treated as a Non-Cited Violation, consistent with
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Appendix C of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as PR 99.9108 (NCV 50-293/99-02-01).
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Conclusions Operators incorrectly moved several control rods during a planned down power due to I
inattention to detail and poor communication. The licensee's immediate corrective l
actions to address this issue were determined to be good. This Severity Level IV
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violation is being treated as a Non-Cited Violation, consistent with Appendix C of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as PR 99.9108 (NCV 50-293/99-02-01).
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Operator Knowledge and Performance O6.1 Main Steam isolation Valve (MSIV) Ooerability issp
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Insoection Scope (71707)
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l The inspector reviewed the operational impact for AO-220-2C, outboard MSIV in the "C"
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steam line, which operated erratically and did not fully close during a slow closure
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surveillance test per 8.7.4.4. This test methodology uses actuator spring force only to slow close the MSIV. Inspector review placed emphasis on the operability evaluation and adherence to technical specification requirements. The remaining seven MSIVs passed the slow closure test.
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Obrervations and Findinos As a result of the test anomaly, operators promptly declared AO-220-2C inoperable and closed the inboard MSIV (i.e., AO-220-1C) in the "C" steam line to meet the i
requirements of TS 3.7.A.2.b, Primary Containment isolation Valves. The inspector l
confirmed that operators followed the guidance in section 7.2 of procedure 2.2.92 which established a leak-off path for condensate to prevent turbine damage and limits reactor power to less than or equal to 75%. Operation with one steam line isolated results in higher reactor pressure in the three other operating lines. The inspector determined that operators followed TS requirements for one inoperable MSIV and the guidance contained in procedure 2.2.92.
Plant management made the decision to operate at reduced power with AO-220-2C inoperable during the remainder of the cycle until RFO12 in May 1999. After consulting with GE Nuclear, the licensee issued a safety evaluation that changed procedure 2.2.92 to allow increasing power up to 90% vice 75%. The inspector confirmed that this higher power level was still bounded by supplemental reload licensing report for Reload 11 Cycle. The licensee increased reactor power to 85%. Two days later a small reactivity and pressure excursion resulted due to instability in the turbine control system. As a precautionary measure, operators lowered power back to 80% where the reactor operated for the remainder of the inspection period. Long term corrective actions were planned to replace the actuator springs on AO-220-2C during RFO1,
The plant experienced previous problems with degraded springs on MSIVs as described in licensee event report (LER) 97-25. Specifically, on November 23,1997, the licensee shutdown for an unplanned outage to resolve test anomalies experienced with two other MSIVs. At that time, the licensee had three sets of springs and used them to replace the closing springs on AO-220-2B,1C and 2D. The licensee ordered five more sets of closing springs for the remaining five MSIVs which are scheduled to be installed during RFO12. Other actions were taken such as installing live load packing and lubricating rollers to enhance smooth operation of the MSIVs.
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Conclusions Operators prorr.ptly declared main steam isolation valve AO-220-2C inoperable due to a slow close test failure and followed technical specifications and other administrative requirements. Subsequently, operators responded well to a small power excursion when operating at 85% reactor power with one steam line isolated.
II. MAINTENANCE M1 Conduct of Maintenance M1.1 General Maintenance a.
Inspection Scooe (61726.62707)
The inspector observed portions of selected surveillance and maintenance activities to verify use of approved procedures, correct system restoration, and proper post work testing. The following activities were observed:
TP99-041 Main Steam isolation Valve Twice Weekly Exercise with MSIV 1C/2C Not Open 8.M.1-32.2 Analog Trip System Trip Calibration - Cabinet C2228-A2 P9400809 Replace 120 Volt EDG Relay MR19900331 Replace SBO EDG cylinder injector tube and injector i
MR19602699 Replace EDG fuel oil transfer system valve 38-HO-129A Top Works b.
Observations and Findinas The inspector verified that the surveillance activities appropriately implemented technical specification (TS) surveillance requirements. Good communication and procedure
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adherence were displayed by the maintenance craft. The pre-job briefing for TP99-041 covered the precautions, prerequisites and expected annunciator alarms.
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Work activities on the station blackout (SBO) EDG included the replacement of one cylinder fuel injector and injector tube. A maintenance mechanic and a lead mechanic performed the work using MR19900331. Also, a maintenance team leader was assigned to the job in an oversight role. The mechanics showed good attention-to-detail by thoroughly cleaning carbon deposits from the top of the cylinder head where the injector nozzle inserts to ensure no leakage.
The work plan specified a particular sequence when tightening fasteners which held the fuel injector into place.' The mechanica noted that the work plan did not contain a specific step at the end to tighten the injector tube fitting to the fuel pump. The mechanics stopped work and sought guidance from the team leader. The decision was made to tighten the fitting hand tight using skills of the trade. After completion of the work, operations ran the SBO EDG to conduct the post work test (PWl') to verify that no fuel oil leakage existed. No extemal leakage from the injector and tube was detected during the PWT.
The inspector identified that a support for the SBO EDG cooling water retum header had a broken fastener. This was a concem because there were only three fasteners securing the support to the engine and since the cooling water header is located above the engine and could be subject to increased vibration. The maintenance team leader informed the inspector that appropriate corrective maintenance actions would be
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initiated.
The inspector questioned the licensee if the replacement fuel injector had been leak j
tested in the shop. The maintenance team leader indicated that the injector had not been shop tested but that any significant intemal leakage and potential adverse effect
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would be detected when running the SBO EDG for the PWT. The inspector verified that cylinder exhaust temperature readings were routinely recorded and evaluated as part of the routinely surveillance test runs. The inspector had no further questions or concems.
The work was completed in a competent manner and successfully passed the PWT.
The inspector observed portions of a planned LCO maintenance outage for the "A" EDG.
The work scope involved multiple activities including the replacement of a DC coil CR120 relay using P9400809 and procedure 3.M.3-55. At the work site, the electricians determined that the replacement relay coil differed from the installed one. Work was stopped and the licensee determined that the replacement relay was an AC CR120 relay rather than a DC CR120 relay. The correct relay was obtained and installed with no further problems. The licensee initiated PR99.0620 to document, evaluate and correct this problem.
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The inspector interviewed the work planner who prepared the package. The planner did not verify that the relay ordered matched the one in the field. A contributing factor to this error was that the majority of other CR120 relays in the plant were of the AC type. The work package was in error by specifying the wrong relay and is the first example of a violation of procedure 1.5.20. This Severity L'evel IV violation is being treated as a
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Non-Cited Violation, consistent with Appendix C of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as PR 99.0620 and PR 99.0636 (NCV 50-293/99-02-02).
Also during the "A" EDG outage, the inspector observed mechanics replacing the top works on fuel oil transfer system valve 38-HO-129A. This valve is a hand operated root valve for a discharge pressure gage. After removing the valve bonnet, the mechanics noticed that the replacement parts were not the same as the ones removed. This was also evident by markings on the valve body which indicated the valve was rated as 3/4 inch with a 600 pound rating. This is the second example of a violation of procedure 1.5.20. (NCV 50-293/99-02-02) The parts that had been ordered were for a % inch with a 800 pound rating. The mechanics informed the maintenance team leader and the
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decision was made to re-install the old parts until the correct parts could be re-ordered.
The licensee initiated PR99.0636 to document, evaluate and correct this problem.
The inspector reviewed the apparent cause evaluation for PR 99.0636 which determined that additional vendor manual (i.e., V-1083) guidance was needed. The inspector determined that this corrective action was adequate. However, the evaluation did not fully discuss the human performance aspects of the planner's actions. For exampie, the parts had been ordered one previous time and were of the wrong size. The planner tried to resolve this verbally with the vendor who incorrectly stated that the 600# rated valves were interchangeable with the 800# rated valves. No problem was initiated during this first event. Additionally, the planner did not research isometric design drawings or contact engineering to determine the proper valve size and rating. The inspector informed the maintenance manager ofthese additional issues not discussed in the apparent cause evaluation. The maintenance manager indicated further review would be performed for these issues.
The inspector was concemed that the two aforementioned work plan quality issues resulted in wrong parts in the field during safety related maintenance, in both cases, maintenance workers in the field identified the wrong parts during planned EDG LCO maintenance outage. The plant production maintenance process owner subsequently initiated a broad review to assess overall work plan quality. Thirty packages were reviewed by the licensee for completeness and quality. Various maintenance groups and quality assurance personnel participated in the review. Of the 30 reviewed,19 had no discrepancies, eight had minor discrepancies and three had problem reports initiated for identified problems. No significant problems were identified. The inspector had no further questions or concems and the broad review was determined to be a good effort to assess and improve overall planning quality.
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Conclusions Overall good pre-job briefs and procedure adherence was displayed during routine maintenance and surveillance activities.
Planned corrective maintenance to replace a fuel injector and high pressure tube on the SBO EDG was successfully completed by competent mechanics. Good attention to l
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detail was evident by thorough ueaning of carbon residue where the injector seals inserts into the cylinder head.
Two different maintenance work package quality issues resulted in wrong parts being sent to the job site for installation in safety related equipment. This was identified by maintenance field workers but reflected errors by planning personnel. In the first example, a work planner error resulted in sending a DC electrical relay coil instead of an AC relay, in the second example, a planner twice ordered the wrong top works for a hand operated valve in the EDG fuel oil transfer system. This Severity Level IV violation is being treated as a Non-Cited Violation, consistent with Appendix C of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as PR 99.0620 and PR 99.0636 (NCV 50-293/99-02-02).
The apparent cause evaluation for the wrong valve parts did not fully evaluate all aspects of how the work planner ordered the wrong parts twice.
M2 Maintenance and Material Condition of Facilities and Equipment M2.1 Refuelino Outaae No.12 (RFO12) Preparations a.
inspection Scope (60705)
A review was performed of the planning and preparations for RFO12 scheduled to begin on May 7,1999. The outage is scheduled to last 30 days with refueling floor activities as the critical path work.
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Observations and Findinas The inspector reviewed the extent that the outage was planned in a manner that minimized shutdown risk. The licensee performed a shutdown risk assessment and issued procedure TP99-057, "RFO Compensatory Measures." Twenty-two issues were identified that had contingency actions developed in advance, For example, due to the extended operational run, the amount of decay heat after shutdown was expected to be higher than previous outages. As a result, both loops of shutdown cooling were planned to remain available while removing the reactor vessel head. The "A" loop of shutdown cooling was planned to be operable while the "B" loop was planned to be available.
Also, major work is scheduled on the 125V DC power panels. During this period a contingency was developed to use a 125V temporary battery via temporary panels to supply critical loads while performing work on the DC pane!s. The inspector determined that the outage safety review was detailed and contained contingency plans to minimize shutdown risk.
The inspector also reviewed the administrative controls established for refueling and several special operations inside the reactor vessel. Procedures were written and contained adequate controls for refueling operations including the removal of a stuck control rod drive mechanism. The inspector noted that the licensee did not plan a full core offload. Two reactor fuel shuffles are scheduled. A detailed safety evaluation was
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completed to ensure that the heat removal capacity of the fuel pool cooling system is not exceeded and remains within the licenchg basis of the plant. The movement of fuel assemblies is scheduled to be performed by coniracted fuel handlers with supervision by a licensed BEC Energy senior reactor operator.
Outage management used performance indicators to track the progress of the development of work packages and modifications. The inspector noted that fewer.
modifications were planned as compared to previous refueling outages. There are 122 modifications planned for RFO12 as compared to 263 completed during RFOf i. Some of the larger projects involve non-safety related equipment such as replacement of the fourth point feed water heater tube bundle. Also, a major modification is scheduled to replace the operators on both feed water regulating valves to allow finer and more reliable control of reactor vessel water level.
The inspector obtained a copy of the quality assurance coverage plan for RFO12. The plan was detailed and emphasized coverage for work that was risk significant. This also included noncafety related equipment such as the feed water regulating valve work and the replacement of many of the feed water heater dump valves.
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Conclusion Extensive planning was evident for RFO12 with an emphasis on minimizing shutdown risk. Adequate administrative controls are in place to control in vessel activities such as repairing a stuck control rod drive and maintaining the reactor fuel offload rate within the capabilities of the decay heat removal systems.
Quality assurance department prepared RFO12 inspection plan partly hated on riGk significance that includes some nonsafety related systams.
M8 Miscellaneous Maintenance issues (92902)
M8.1 (Closed) LER 50-293/98-25: Setooint of f amet Rock Relief Valve Found Out of Tolerance Durina Testina This LER documented that the as-found popping pressure of an installed target rock main steam relief valve was greater than the technical specification limit of 1115 +/- 11 psig. The relief valve lifted at 1141 psig. The cause is believed to be set point drift.
Problem report 98.9593 was written to document this condition.
The inspector conducted an on-site review of the LER and reviewed the licensee's proposed corrective actions. The valve as tested was within +/- 3% tolerance and is not presently installed in the plant. The lift pressure for the relief was sufficient to prevent over pressurization of the nuclear steam system during the limiting over pressurization event. No violation of NRC requirements or new concems were identifed by the inspector. This LER is closed.
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M8.2 (Closed) LER 50-293/98-26: Hiah Pressure Coolant Iniection (HPCI) System Declared japporable Due to Failed Power Inverter This LER documented that the inverter which supplies the HPCl flow control circuitry tripped and failed to automatically restart. Problem report 98.9604 was initiated to document the event and determine the root cause. As corrective action, the licensee replaced the inverter and sent the inverter to the manufacture for further evaluation.
The inspector conducted an on-site review of the LER and verified that the inverter had been replaced and that the event was captured in the licensee's corrective action program. No violation of NRC requirements or new concems were identified by the inspector. This LER is closed.
M8.3 (Closed) LER 50-293/98-27: Switchaear Room Door inooerable for Tomado Analysis This LER documented that the door to the 4.16Kv switchgear room "B" was inoperable due to degraded door hinges. Problem report 98.9634 was initiated and the door repaired. The problem was found during routine preventive maintenance.
The inspector conducted an on-site review of the LER and verified that the door was repaired and that the licensee has a procedure in place to inspect this door monthly. No violation of NRC requirements or new concems were identified by the inspector. This LER is closed.
l lil. ENGINEERING E2 Engineering Support of Facilities and Equipment E2.1 Plant Desian Chance Review a.
Insoection Scooe (37551)
l The inspector reviewed selected safety-related plant design change (PDC) records I
implemented during the cycle 11 refueling outage (RFO) or scheduled to be implement during RFO 12, scheduled for May 8,1999. The purpose of the review was to verify that the PDCs were implemented in accordance with NRC requirements and licensee procedures NE3.20," Preparation, Review, Approval, Revision, and Closeout of Modifications at Pilgrim Station," and NOP83E1, " Control of Modifications it Pilgrim Station."
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Observations and Findinas The inspector reviewed the following four PDC records; two of which are scheduled to be implemented during the current refueling outage and two that were incorporated curing RFO 11.
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PDC 96-32, Residual Heat Removal and Core Spray Pump Suction
Strainer Replacement PDC 97-06, Containment Spray Nozzle Modification
PDC 98-21, Motor Operated Valve Modifications
PDC 98-23, Replacement Of Outlier Relays
For the PDCs installed during RFO 11, the inspector verified that operations critical drawings were updated, required pre-operational training complete, applicable procedures updated and other required administrative items were completed prior to declaring the applicable system / component operable. For the modifications scheduled to be implemented duiing RFO 12, the inspector verified that items that are required to be completed prior to releasing the system for system operability were clearly defined.
The inspector also assessed the basis of the plant modifications, the technical evaluation scope, and safety evaluation conclusions. All of the PDCs were reviewed and approved by the plant operation review committee. No major problems were identified.
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Conclusions The PDCs reviewed clearly identified those controlled documents and training activities that were required to be updated prior to tumover of the component / system to operation.
The PDCs were properly implemented in accordance with NRC and licensee requirements.
IV. PLANT SUPPORT R1 Radiological Protection and Chemistry (RP&C) Controls During entry to and exit from radiologically controlled areas (RCA), the inspectors verified that proper wamings were posted, personnel entering were wearing proper dosimetry, and personnel and materials leaving the RCA were properly monitored for radioactive contamination. The inspectors also verified that access to locked high radiation areas was properly being controlled.
S1 Conduct of Security and Safeguards Activities The inspector toured the protected and vital areas and verified that personnel were properly badged for unescorted or escorted access. Inspection of the protected area barrier revealed that protective area gates were locked or guarded and isolation zones were free of obstructions. The inspector also verified that the temporary structures installed in preparation for the cycle 12 refueling outage were properiy illuminated for security precaution e I
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V. MANAGEMENT MEETINGS X1 Exit Meeting Summary i
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i The inspector met with the licensee representatives at the conclusion of the inspection I
on May 17,1999. At that time, the purpose and scope of the inspection were reviewed, and the preliminary findings were presented. The licensee acknowledged the j
preliminary findings.
J X3 Management Meeting Summary On April 15,1999, a working level meeting was held in NRC Region 1 office between the licensee and NRC Region 1 personnel to discuss the number of open operability evaluations.
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ATTACHMENT 1
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INSPECTION PROCEDURES USED IP 37551:
Onsite Engineering IP 40500:
Effectiveness of Licensee Controls in Identifying, Resolving, and Preventing Problems IP 61726:
Surveillance Observation IP 62707:
Maintenance Observation IP 71707:
Plant Operations IP 71750:
Plant Support Activities IP 82301:
Evaluation of Exercises for Power Reactors IP 92700:
Onsite Followup of Written Reports of Nonroutine Events at Power Reactor Facilities IP 92901:
Followup - Operations IP 92902:
Followup - Maintenance IP 92903:
Followup-Engineering IP 92904:
Followup - Plant Support IP 93702:
Prompt Onsite Response to Events at Operating Power Reactors l
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l-I Attachment 1
O ITEMS OPENED, CLOSED, AND UPDATED
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Closed LER 50-293/98-21 '
Inadequate Fuel Supply for Emergency Diesel Generators (EDGs)
l LER 50-293/98-23 Incorrect Wiring Modifications Affected Reactor Building Closed l
- Cooling Water (RBCCW) Train "B" Alternate Shutdown Panel l
LER 50-293/98-25 Setpoint of Target Rock Relief Valve Found Out of Tolerance During l
Testing
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LER 50-293/98-26 High Pressure Coolant injection (HPCI) System Declared inoperable Due to Failed Power inverter LER 50-293/98-27 Switchgear Room Door inoperable for Tornado Analysis i
NCV 50-293/99-02-01 Failure to insert control rods in accordance with the rod withdrawal i
sheets NCV 50-293/99-02-02 Work package errors (two examples)
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Attachment 1
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LIST OF ACRONYMS USED BECo Boston Edison Company CFR Code of Federal Regulations CRHEAF Control Room High Efficiency Air Filtration DRP Division of Reactor Projects EDG Emergency Diesel Generator FSAR Final Safety Analysis Report IFl inspection Follow-Up item IR
Inspection Report
LCO
Limiting Condition of Operation
LER
Licensee Event Report
Maintenance Request
Non-Cited Violation
Nuclear Plant Operator
NRC
Nuclear Regulatory Commission
Office of Nuclear Reactor Regulation
Public Document Room
Pilgrim Nuclear Power Station
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PR
Problem Report
Post Work Test
Radiologically Controlled Areas
Refueling Outage
Radiological Protection
SBLC
Station Blackout
Updated Final Safety Analysis Report
Violation
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